ML20209B842

Forwards Revised Draft Version of Section 11 of NUMARC 93-01, Assessment of Risk Resulting from Performance of Maint Activities, for Info.Public Meeting Scheduled & Noticed for 990709 to Discuss Revised Section 11

ML20209B842
Person / Time
Issue date:	07/07/1999
From:	Scott W NRC (Affiliation Not Assigned)
To:	Quay T NRC (Affiliation Not Assigned)
References
NUDOCS 9907080183
Download: ML20209B842 (15)

Text

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y July 7,1999 MEMORANDUM TO:

Theodore R. Quay, Chief Quality Assurance, Vendor inspection, Maintenance, and Allegations Branch Original signed by; Wayne E. Scott, Jr.

FROM:

Wayne Scott Reliability and Maintenance Section Quality Assurance, Vendor inspection, Maintenance, and Allegations Branch

SUBJECT:

REVISED DRAFT OF CHAPTER 11 OF NUMARC 93-01 i

On July 1,1999, the Nuclear Energy hstitute (NEI) forwarded to the NRC a revised (undated) draft version of Section 11 of NUMARC 93-01, " Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." The revised Section 11, titled

" Assessment of Risk Resulting from Performance of Maintenance Activales,"is attached for your information. It was prepared in response to the recent revision of 10 CFR 50.65, the maintenance rule, which added a requirement to assess and manage the risk resulting from l

nuclear power plant maintenance activities and which was addressed by SECY 99-133, dated May 17,1999.

l A public meeting has been scheduled and noticed for July 9,1999, to discuss the revised Section 11 of NUMARC 93-01 l

Attachment:

As stated l

Distribution:

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1.0 ASSESSMENT

OF RISK RESULTING FROM PERFORMANCE OF MAINTENANCE l

ACTIVITIES i

11.1 Reference l

10 CFR 50.65(a)(4) l Before performing maintenance activities (including but not limited to surveillance, post maintenance testing, corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. The scope of structures, systems, and components (SSCs) to be included in the assessment may be limited to those SSCs that a risk-informed evaluation process has shown to be significant to public health and safety.

11.2 Background i

Maintenance activities must be performed to provide the level of plant equipment reliability necessary for safety, and should be carefully managed to achieve a balance between the benefits and potential impacts on safety, reliability and availability.

u The benefits of well managed maintenance conducted during power operations include increased system and unit availability, reduction of equipment and system deficiencies that could impact operations, more focused attention during periods when fewer activities are competing for specialized resources, and reduction of work scope during outages. In addition, many maintenance activities may be performed during power operation with a smaller net risk impact than during outage conditions, particularly for systems whose performance is most important during shutdown, or for which greater functional redundancy is available during power operations.

11.3 Guidance This section provides guidance for the development of an approach to assess and manage the risk impact expected to result from performance of maintenance activities. Assessing the risk means using a risk-informed process to evaluate the overall contribution to risk of the planned maintenance activities. Managing the risk means providing proper awareness of the risk to plant personnel, and taking actions as appropriate to control the risk.

The assessment is required for maintenance activities perfomied during power operations or during shutdown. Performance of maintenance during power operations should be planned and scheduled to properly control out-of-service time of systems or equipment. Performance of maintenance activities during shutdown should consider the impact on performance of key shutdown safety functions.

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11.3.1 Assessment Process, Control, and Responsibilities l

The process for conducting the assessment and using the result of the assessment in plant decisionmaking should be proceduralized. The procedures should denote l

responsibilities for conduct and use of the assessment, and should specify the plant functional organizations and personnel involved, including, as appropriate, operations, engineering, and risk assessment (PSA) personnel.

I1.3.2 General Guidance for the Assessment - Power Operations and Shutdown

1. Power Operating conditions are defined as plant modes other than hot shutdown, cold shutdown, refueling, or defueled. Section 11.2.3 describes the scope of SSCs subject to the assessment during power operations. Section 11.2.5 describes the scope of SSCs subject to the assessment during shutdown.

2. The assessment' method may use quantitative approaches, qualitative approaches, or blended methods. In general, the assessment shocid consider:

The degree of redundancy available for performance of the safety function served by the out-of-service SSC The duration of the out-of-service condition The likelihood of the need for performance of the affected safety function to mitigate an initiating event or accident The likelihood that the maintenance activity willincrease the frequency of an initiating event.

Component and system dependencies that are affected.

3. The assessments may be predetermined or performed on an as-needed basis.

4. The degree of depth and rigor used in assessing and managing risk should be related to the safety significance of the SSCs planned for maintenance, and the impact of the maintenance activity on the train or system function.

5. The assessment should take into account whether the out-of-service SSCs could be promptly restored to service if the need arose due to emergent conditions. The assessment should consider the time necessary for restoration with respect to the time at which performance ofits safety function would be needed. Examples include:

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. An SSC out-of-service for monitoring or surveillance might be capable of prompt restoration of function, whereas an SSC out of service for extensive maintenance would typically not be promptly available.

An SSC 'placed in pull-to-lock (automatic feature disabled) would still be l

functional in that the system would start if switched to run.

6. Emergent conditions may arise following original performance of an assessment, and before or during the maintenance activity, that could affect the assessment results. Examples include plant configuration or mode changes, additional SSCs out of service due to failures, or significant changes in extemal conditions J(weather, offsite power availability). The following guidance applies to this situation:

The safety assessment should be re-evaluated to address the changed plant conditions on a reasonable schedule commensurate with the safety significance of the condition. Based on the results of the re-evaluation, ongoing or planned maintenance activities may need to be suspended or rescheduled, and SSCs may need to be retumed to service, Re-evaluation of the assessment should not interfere with, or delay, the e

operator and/or maintenance crew from taking timely actions to restore the equipment to service or take compensatory actions.

. If the previously assessed plant configuration is restored to service prior to re-i 1

evaluating the assessment, the assessment need not be re-evaluated.

11.3.3 Scope of Assessment for Power Operating Conditions The scope of the Systems, Structures and Components (SSCs) to be addressed by the assessment for power operating conditions is as follows:

1. Those SSCs included in the scope of the plant's level one, intemal events probabilistic safety assessment (PSA), and;.

' 2; SSCs in addition to the above that have been determined to be high safety significant (risk significant) through the process described in Section 9.3 of this

' document.

1 With regard to item 1 above, the PSA used as the basis for determining the scope of the (a)(4)~ assessment should include consideration of support systems and dependencies for SSCs that impact plant risk. The industry PSA peer review process (Reference NEl-99-XX) includes information for evaluation of the correct treatment of these attributes in developing a PSA. This guidance is located in Appendix B of the

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~ peer review guidelines. Section SY discusses systems analysis and treatment of support systems, and Section DE discusses treatment of dependencies.

SSCs within the above scope rnay be eliminated from further consideration for the (a)(4) assessments if they are evaluated and shown to have little or no impact on key plant safety functions.

11.3.4 Assessment Methods for Power Operating Conditions

1. _ A single SSC may be rernoved from service for a reasonable amount of time (as 1

constrained by Technical Specifications, if pertinent to the SSC), without conduct of

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the assessment, unless unusual conditions are present or imminent (e.g., severe l

weather, offsite power instability).

2. The assessment should be conducted prior to planned maintenance activities that simultaneously affect more than one SSC.

As noted, the assessment may be performed using quantitative, qualitative, or blended methods. Additional guidance is provided below:

11.3.4.1.' Quantitative Considerations For maintenance resulting in simultaneous removal from service of multiple SSCs

within the scope of the PSA, the assessment process may be informed by a tool or u

method.that considers quantitative insights from the PSA. This can take the form of using the PSA model, or using a safety monitor, matrix, or list derived from the PSA insights.R l

The PSA model should include the following characteristics, or, if not, it's limitations for use in supporting the assessment should be compensated for by additional qualitative evaluation. The EPRI PSA Applications Guide (EPRI TR-105396) discusses considerations regarding PSA attributes, maintenance, and use in decisionmaking. This guidance should be considered in determining the degree of confidence that can be placed in the use of the PSA for the assessment, and whether additional qualitative considerations should be brought to bear:

. ~ The PSA ' hould address intemal initiating events (exclusive of intemal fires) s L

e The PSA should provide level one insights (contribution to core damage frequency).

The PSA should be reviewed periodically and updated as necessary to provide reasonable representation of the current plant design.

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l The PSA may model the plant at the system, train, or component level, or l

l combinations of the above.

For the purpose of the assessment, the existing PSA is not required to be expanded to quantitatively address containment performance (level 2), extemal events, or conditions other than power operation. Strengthening of the existing PSA represents an optional approach to support conduct of the assessment.

l If the PSA is modeled at the train or system level, the assessment should include i

consideration of the impact of the out of service SSC on the safety function of the train or system. SSCs are considered to support the train or system if the SSC is a

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Esignificant component of the success path for function of the train or system (e.g.,

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primary pump, or valve in primary flowpath). However, if the SSC removed from l

service does not contribute significantly to the train or system safety function (e.g.,

indicator light, alarm, drain valve), the SSC is not considered to support the system or train.

If the PSA is modeled at the system level and the removal from service of an SSC resuhs in the unavailability of one train of a multi-train system, the assessment should consider this impact, and should not assume that full function of the system is maintained.

11.3.4.2 Qualitative Consideratior 3 l

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1. The assessment may be perfonned by a qualitative approach, by addressing the impact of the maintenance activity upon key safety functions, as follows:

l identify key safety functions affected by the SSC planned for removal from e

service.

j Consider the degree to which removing the SSC from service will impact the key safety functions.

i Consider degree of redundancy, duration of out-of-service condition, and e

appropriate compensatory measures, contingencies, or protective actions that can be taken.

2. For power operation, key plant safety functions are those that ensure the integrity i

of the reactor coolant pressure boundary, ensure the capability to shut down and maintain the reactor in a safe shutdown condition, and ensure the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposure comparable to 10 CFR Part 100.

Examples of these power operation key safety functions are:

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. for PDR Containment integrity (Containment isolation, Containment Pressure and Temperature Control);

Reactivity Control; e

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. - Reactor Coolant Heat Removal; and l-Reactor Coolant Inventory Control.

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3. The key safety functions are achieved by using systems or combinations of systerns, that could include redundant subsystems or trains. For the purposes of 1

the equipment out-of-service assessment, SSCs are considered to support a key safety function if they:

l Have a significant impact on the performance of a key safety function; or e

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Have a significant potential to challenge a key safety function, such as SSCs whose failure would result in a scram or safety system actuation, or would significantly complicate recovery efforts.

' 4. The analysis should include the assessment of plant systems supporting the affected key safety functions, and trains supporting these plant systems 5; Qualitative considerations may also be necessary to address extemal events, containment performance issues, and SSCs not in the scope of the level one, intemal events PSA (e.g., included in the assessment scope because of expert

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panel considerations). In these cases, the assessment may need to include j

consideration of actions which could affect the ability of the containment to perform its function as a fission product barrier. With regard to containment performance, the assessment should consider:

Whether new containment bypass conditions are created, or the probability of containment bypass conditions is increased;

.. - Whether new containment penetration failures that can lead to loss of l

containment isolation are created; and.

. If maintenance is performed on components of the containment heat removal system, whether redundant containment heat removal trains should be available.'

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6. Extemal event considerations involve the potential impacts of weather, flooding, and seismic conditions with regard to the proposed maintenance evolution. For

- the purposes of the assessment, weather and external flooding need to be considered if such conditions are imminent or have a high probability of occurring during the planned out-of service duration. An example where these

-- considerations are appropriate would be the long term removal of exterior doors or floor plugs. Intemal flood issues are associated with the potential for a condition to exist during a planned maintenance evolution when one maintenance activity poses an increased risk of flood and a second activity exposes to that flood hazard those SSCs needed to perform key safety functions.

11.3.5 ' Scope of Assessment for Shutdown Conditions The scope of the Systems, Structures and Components (SSCs) to be addressed by the assessment for shutdown conditions are those SSCs necessary to support the following shutdown key safety functions (from Section 4 of NUMARC 91-06):

. Decay heat removal capability

. Inventory Control-e' PowerAvailability Reactivity control '

= Containment (primary / secondary)

The shutdown key safety functions are achieved by using systems or combinations of

. systems, that may include redundant subsystems or trains. The shutdown assessment need not be performed for SSCs whose operability is not required by Technical Specifications during shutdown mode, unless these SSCs are considered for establishment of backup success paths or compensatory measures 11.3.6 Assessment Methods for Shutdown Conditions NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management, Section 4.0, provides a complete discussion of shutdown safety considerations with respect to maintaining key shutdown safety functions, and should be considered in developing an assessment process that meets the requirements of 10 CFR 50.65(a)(4).

. Perfomiance of the safety assessment for shutdown conditions generally involves a qualitative assessment with regard to key safety functions, and follows the same

general process described in Section 11.2.4.2.3 above. (Those plants that have perfomled shutdown PSAs can use these PSAs as the basis for their shutdown

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' for PDR assessment methods.) However, there are some different considerations than the at-power assessment. These include:

The shutdown assessment is typically focused on SSCs "available to perfonn a function" versus SSCs "out of service"in the case of power operations. Due to decreased equipment redundancies during outage conditions, the outage planning and control process involves consideration of contingencies and backup methods to achieve the key safety functions, as well as on measures that can reduce both l

the likelihood and consequences of adverse events.

Assessments for shutdown maintenance activities need to take into account plant conditions and multiple plant configurations that impact the shutdown key safety functions. The shutdown assessment is a component of an effective outage planning and control process Maintenance activities that do not necessarily remove the SSC from service may stillimpact plant configuration and impact key safety functions. Examples include:

A valve manipulation that involves the potential for a single failure to create a draindown path affecting the inventory control key safety function A switchyard circuit breaker operation that involves the potential for a single failure to he affects availability of AC power.

Because of the special considerations of shutdown assessments, additional guidance is provided below with respect to each key safety function:

.I1.3.6.1 Decay Heat Removal Capability Assessments for maintenance activities affecting the DHR system should consider that other systems and components can be used to remove decay heat depending on a variety of factors, including the plant configuration, availability of other key safety systems and components, and the ability of operators to diagnose and respond properly to the event. For example, assessment of maintenance activities that impact the decay heat removal key safety function should consider:

Initial magnitude of decay heat time to boiling time to core uncovery time to containment closure e

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for PDR Initial RCS water inventory condition (e.g., filled, reduced, mid-loop, refueling canal filled, reactor cavity flooded, etc.)

i RCS configurations (e.g, open/ closed, nozzle dams installed or loop isolation e'

valves closed, steam generator manways on/off, vent paths available, temporary

- covers or thimble tube plugs installed, main steam line plugs installed, etc.)

. - natural circulation capability with heat transfer to steam generator shell side

'If the fuelis offloaded to the spent fuel pool during the refueling outage, the decay heat removal function is shifted from the RCS to the spent fuel pool. Assessments for maintenance activities should reflect appropriate planning and contingencies to

- address loss of SFP cooling.

I1.3.6.2 -

Inventory Control Assessments for maintenance activities should address the potential for creating inventory loss flowpaths. For example, Maintenance activities associated with the main steam lines (e.g., safety / relief valve removal, automatic depressurization system testing, main steam isolation valve maintenance, etc.) can create a drain down path for the reactor cavity and fuel pool.

For BWRs, there are potential inventory loss paths through the DHR system to the suppression pool when DHR is aligned for shutdown cooling

- Assessments for maintenance activities during reduced inventory operations are especially important. Reduced inventory operation occurs when the water level in the reactor vessel is lower than 3 feet below the reactor vessel flange. A special case of j

reduced inventory operation for PWRs is mid loop operation, which occurs when the RCS water level is below the top of the hot legs at their junction with the reactor vessel. Similar conditions can exist when the reactor vesselis isolated from steam generators by closed loop isolation valves or nozzle dams with the reactor vessel head installed or prior to filling the reactor cavity. Upon loss of DHR under these conditions, coolant boiling and core uncovery can occur if decay heat removal is not restored or j

provided by some altemate means. In addition, during mid-loop operation, DHR can be lost by poor RCS level control or by an increase in DHR flow (either of which can j

' ingest' air into the DHR pump).

11.3.6.3 Power Availability _

Assessments should consider the impact of maintenance activities on availability of electrical power.1 Electrical power is required during shutdown conditions to maintain

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Assessments for maintenance activities involving AC power sources and distribution systems should address providing defense in depth that is commensurate with the plant condition.

Assessments for maintenance activities involving the switchyard and transformer yard should consider the impact on offsite power availability.

AC and DC instrumentation and control power is required to support systems that provide key safety functions during shutdown. As such, maintenance activities affecting power sources, inverters, or distribution systems should consider their functionality as an important element in providing appropriate defense in depth.

I1.3.6.4 Reactivity Control a

l The main aspect of this key safety function involves maintaining adequate shutdown margin in the RCS and the spent fuel pool. Maintenance activities involving addition of water to the RCS or the refueling water storage tank have the potential to result in l

Boron dilution. During periods of cold weather, RCS temperatures can also decrease below the minimum value assumed in the shutdown margin calculation.

I1.3.6.5 Containment - Primary (PWR)/ Secondary (BWR)

Maintenance activities involving the need for open containment should include evaluation of the capability to achieve containment closure in sufficient time to mitigate potential fission product release. This time is dependent on a number of factors, including the decay heat level and the amount of RCS inventory available.

In addition to the guidance in NUMARC 91-06, for plants which obtain license amendments to utilize shutdown safety administrative controls in lieu of Technical Specification requirements on primary or secondary containment operability and ventilation system operability during fuel handling or core alterations, the following guidelines should be included in the assessment of systems removed from service:

During fuel handling / core alterations, ventilation system and radiation monitor availability (as defined in NUMARC 91-06) should be assessed, with respect to filtration and monitoring of releases from the fuel. Following shutdown, radioactivity in the fuel decays away fairly rapidly. The basis of the Technical Specification operability amendment is the reduction in doses due to such decay. The goal of maintaining ventilation system and radiation monitor 7/1/99 10 l

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' availability is to reduce doses even further below that provided by the natural L

decay, and to avoid unmonitored releases.

j A single normal or contingency method to promptly close primary or secondary containment penetrations should be developed. Such prompt methods need not completely block the penetration or be capable of resisting pressure. The purpose is to enable ventilation systems to draw the release from a postulated fuel handling accident in the proper direction such that it can be treate' and-d monitored.

I1.3.7 Managing Risk l

The assessment provides insights relative to the risk-significance of maintenance activities. The process for managing risk related to maintenance activities involves'using the result of the assessment in plant decisionmaking to control the risk increases. This is accomplished through pmdent planning, scheduling, coordinating, monitoring, and adjusting maintenance activities to manage the risk impact. This process should include an understanding of the nature (e.g., affecting the key safety function, the core damage, or large early release frequency) and associated actions based on that understanding.

l The risk impact of maintenance activities may be controlled by defining i

appropriate action hvels. These actions can include operating shift personnel awareness, plant management approval, and establishment of compensatory actions, contingency plans, or altemate success paths.

The effective control of risk increases due to an unexpected failure of a risk-

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important SSC can be reasonably assured by planning for contingencies, or coordinating, scheduling, monitoring, and modifying the duration of planned maintenance activities.

' For maintenance configurations involving higher risk for a short duration, the e

duration of the me itenance activity should be minimized through appropriate planning and preparation, such as briefings, training on mockups, and prestaging necessary materials and equipment.

~. Appropriate site personnel should be at a heightened state of risk awareness while the plant is in the configuration.

I1.3.7.1 Action levels The process for management of risk should include establishment oflevels for actions as discussed above. The following factors should be considered in establishing the action levels:

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DRAFTof 7/1/99 for PDR The remaining mitigation capability, e.g., the degree of redundancy available for performance of the safety function served by the out-of-service SSC, including consideration of compensatory measures and contingencies. A greater degree of redundancy should result in a lower risk impact The duration of the out-of-service condition. A shorter duration should result in a smaller risk impact.

The expected frequency of the initiating event for which the performance of the safety function would be required. A lower frequency of initiators should result in a lower risk impact. Examples are shown below:

3 Approximate frequency Example event type 1-10 years reactor trip, loss of main feedwater, loss of condenser 10-100 years LOOP, SGTR, stuck open SRV (BWR), MSLB (outside containment), loss of I SR bus, loss of instrument air, fire 100-1,000 years small LOCA (PWR), stuck open PORV/SRV, MFLB, flood 1,000-10,000 years med LOCA (PWR), small LOCA (BWR), MSLB inside containment, loss of all service water 10,000-100,000 years large/ medium LOCA (BWR)

Greater than 100,000 years large LOCA (PWR), ISLOCA, vessel rupture, severe earthquake Other methods for establishing action levels may include use of quantitative insights from the PSA. A number of acceptable approaches exist, and may be used singularly or in combination.

The baseline risk level from which the risk increase is assessed may be the standard annual baseline risk level (incorporating the contribution to risk of equipment out-of-service due to maintenance).

The baseline risk level from which the risk increase is assessed may be the "zero maintenance" model, which corresponds to a condition where all equipment is in service and the only contribution to risk is the random failure rates for components and operators and random initiating event frequencies.

The action level may include consideration of a specific value of the CDF (or LERF, if calculated) that results from the maintenance activity. This value may be defined as an absolute risk level, or as a relative increase to one of the baseline levels discussed above.

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.. g The action level may include consideration of the incremental risk increases due to individual maintenance activities over a set time period. This approach involves consideration of the integrated risk incurred over a period of time a' configuration or condition exists, and can be expressed as core damage probability (CDP) or large early release frequency probability (LERP).

The action level criteria may include consideration of a cumulative risk value, based 2 e on computing the total cumulative risk due to maintenance activities over a specific interval.

Due'to differences in plant type and design, there is acknowledged variability in baseline core damage frequency. - Further, there is variability in containment perforrnance that may impact the relationship between baseline core damage frequency and baseline large early release frequency for a given plant or class of plants.

-Therefore, determination of the appropriate method or combination of methods as discussed above, and the corresponding quantitative decision criteria, are plant-unique j

activities.

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11.2.8 Documentation-The following are guidelines for documentation of the safety assessment:

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1. The purpose of this section of the maintenance rule is to establish a process for assessment ofimpacts on key safety functions due to planned removal of equipment from service. This purpose should be effected through establishment of plant procedures that address process, responsibilities, and decision approach. it may also be appropriate to include a reference to the appropriate procedures that govem planning and scheduling of maintenance or outage activities. The process itself should be documented.

2. The normal work control process should suffice as a record that the assessment was performed. It is not necessary to document the basis of each assessment for removal of equipment from service as long as the process is followed. For evaluation of removal from service of multiple SSCs using a predetermined l

approach (such as a safety monitor, list, or matrix), no further documentation is necessary unless additional special considerations (such as compensatory measures, or consideration ofissues beyond the scope of the assessment tool) l are involved.

3. In special' situations where the normal assessment tools may be unavailable or not applicable, it may be necessary to rely on operator judgment as the basis for the assessment. This situation should be addressed by the proceduralized process above.

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Definitions I

i Current definition of Unavailability:

The numerical complement of availability. An SSC that cannot perform its intend function. An SSC that is required to be available for automatic operation must be available and respond without human action.

Proposed definition of Unavailability Unavailability = olanned unavailable hours + unnianned unavailable hours required hours Availability: An SSC is available ifit is capable of performing its intended function under realistic conditions. A SSC is still considered available if it can be restored to functional status within the timeframe associated with realistic conditions using proceduralized operator actions. Faulted exposure time should be accounted for using reliability - not be included as a term in unavailability.

(Realistic conditions are "best estimate" conditions assumed in establishing PSA i

success criteria which includes credible design basis accident conditions.)

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